Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 4, 1999, through December 17, 1999.

The last biweekly notice was published on December 15, 1999 (64 FR

70077).

Notice of Consideration of Issuance of Amendments to Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

Involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By January 28, 2000, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC, and electronically from

the ADAMS Public Library component on the NRC Web site, http://

www.nrc.gov (the Electronic Reading Room). If a request for a hearing

or petition for leave to intervene is filed by the above date, the

Commission or an Atomic Safety and Licensing Board, designated by the

Commission or by the Chairman of the Atomic Safety and Licensing Board

Panel, will rule on the request and/or petition; and the Secretary or

the designated Atomic Safety and Licensing Board will issue a notice of

a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Docketing and

Services Branch, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

by the above date. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions,

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supplemental petitions and/or requests for a hearing will not be

entertained absent a determination by the Commission, the presiding

officer or the Atomic Safety and Licensing Board that the petition and/

or request should be granted based upon a balancing of factors

specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of amendments request: November 22, 1999.

Description of amendments request: The proposed amendment revises

Technical Specification (TS) 5.5.11, ``Ventilation Filter Testing

Program'' for laboratory testing of charcoal in Clavert Cliffs

engineered safety feature (ESF) ventilation systems to reference the

latest charcoal testing standard (American Society for Testing and

Materials [ASTM] D3803-1989, ``Standard Test Method for Nuclear-Grade

Activated Carbon''). This TS change was requested by the Nuclear

Regulatory Commission (NRC) in Generic Letter 99-02, ``Laboratory

Testing of Nuclear-Grade Activated Charcoal,'' and is based on the

NRC's determination that testing nuclear-grade activated charcoal to

standards other than ASTM D3803-1989 does not provide assurance for

complying with the current licensing basis as it relates to the dose

limits of General Design Criterion 19 of Appendix A to Part 50 of Title

10 of the Code of Federal Regulations (10 CFR) and Subpart A of 10 CFR

Part 100. The generic letter provided a sample TS that the NRC

considers acceptable. The proposed revision to TS 5.5.11 meets the

intent of the sample TS. Specifically, the proposed change removes the

reference to testing in accordance with American National Standards

Institute N510-1975 and changes the allowable methyl iodide penetration

to an acceptance criterion that is derived from applying a safety

factor of two to the charcoal filter efficiency assumed in Calvert

Cliffs design basis dose analysis. The proposed changes will ensure

that the charcoal filters used in ESF ventilation systems will perform

in a manner that is consistent with the particular ESF charcoal

adsorption efficiencies assumed in the analyses of design basis

accidents.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

This proposed change makes changes to the methods, test

conditions, and acceptance criteria associated with the performance

of the laboratory tests of charcoal samples. The effected equipment

is used to mitigate the consequences of an accident and are not

accident initiators. This proposed change does not make any changes

to the method of obtaining the charcoal sample. No structural

changes or modifications are being made to the ESF ventilation

equipment. This proposed change does not make any changes to

equipment, procedures, or processes that increase the likelihood of

an accident. Therefore, this proposed change does not involve a

significant increase in the probability of an accident previously

evaluated.

The ESF ventilation systems are designed to mitigate the

consequences of accidents. The design basis analysis of the

accidents account to varying degrees for the reduction in airborne

radioactive material provided by the charcoal filters. The proposed

change will change the charcoal filter test protocol to ASTM D3803-

1989. The use of this standard will produce more accurate and

reproducible laboratory test results and provides a more

conservative estimate of charcoal filter capability. The proposed

change makes changes to the methyl iodide penetration acceptance

criteria to ensure that the charcoal filters are capable of

performing their required safety function for the expected operating

cycle. The proposed change will make it more likely that the

charcoal will meet its intended safety function as described in the

Updated Final Safety Analysis Report. Therefore, the proposed change

does not significantly increase the consequences of an accident

previously evaluated.

Based on the above, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The proposed change will not make any physical changes to the

plant or changes to the ESF ventilation system operation. The

proposed change is limited to the ESF ventilation system testing

protocol, test conditions, and acceptance criteria. These changes

are administrative in nature. This proposed change does not make any

changes to the method of obtaining the charcoal sample. This

proposed change does not cause any ESF ventilation equipment to be

operated in a new or different manner. No structural changes or

modifications are being made to the ESF ventilation equipment. This

proposed change does not create any new interactions between any

plant components. Therefore, the possibility of a new or different

type of accident is not created by this proposed change.

3. Would not involve a significant reduction in a margin of

safety.

The safety function of the ESF ventilation systems is to

mitigate the consequences of accidents by reducing the potential

release of radioactive material to the environment or the Control

Room following a design basis accident. The TS requirements for

laboratory testing of charcoal samples provides assurance that the

charcoal filters in these systems are capable of reducing airborne

radioactive material to within acceptable limits. The proposed

license amendment requires the use of the latest NRC-accepted

charcoal testing standard and makes changes to the charcoal testing

methyl iodide removal efficiency acceptance limits in accordance

with the formula provided by the NRC in Generic Letter 99-02. The

proposed license amendment continues to provide assurance that the

charcoal filters are capable of reducing airborne radioactive

material to within acceptable limits. Therefore, the proposed change

does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Sheri R. Peterson.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of amendments request: November 22, 1999.

Description of amendments request: The Baltimore Gas and Electric

Company (BGE) requests an amendment to implement a change to the

Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety

Analysis Report (UFSAR) that constitutes an unreviewed safety question

as described in 10 CFR 50.59.

The change revises the information currently provided within the

UFSAR on aircraft and their flight paths for Patuxent River Naval Air

Station (Pax River NAS). The existing information is outdated and does

not reflect current conditions for aircraft utilizing Pax River NAS.

Additionally, the UFSAR will be revised to add information

[[Page 73086]]

pertaining to the corporate helipad located northwest of the plant.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The probability of an aircraft crash was not quantified during

the timeframe of licensing and construction of the plant. As was

noted previously, the Directorate of Licensing at the U.S. Atomic

Energy Commission concurred with Baltimore Gas and Electric

Company's conclusion that no special design provisions were required

to be incorporated into Calvert Cliffs Nuclear Power Plant (CCNPP)

because the probability of an aircraft crash affecting the plant was

acceptably low (implies a probability of less than 10-7/

Year). Therefore, the probability of an aircraft crash affecting the

plant was acceptably low at less than 10-7/year.

The probability of an aircraft accident resulting in

radiological consequences greater than 10 CFR Part 100 exposure

guidelines was considered to still be below the Standard Review Plan

(SRP) (NUREG-0800) level of acceptability of 1.0 x 10-7

per year for CCNPP. The probability of an aircraft accident during

the timeframe of original construction and licensing of the plant

was never quantified. Since today's probability of an aircraft

accident may be higher based on the fact that, at times, aircraft

going into Patuxent River Naval Air Station fly over the plant,

where previously they came no closer than seven miles from the plant

(as described in the UFSAR), the probability of occurrence of an

accident will conservatively be considered to have increased.

However, it should be noted that the probability of an aircraft

accident resulting in radiological consequences greater than 10 CFR

Part 100 exposure guidelines is still considered to be below

1.0 x 10-7 cr/yr, which is acceptable since it is within

SRP Section 3.5.1.6 guidelines. Since the above probability of an

aircraft accident meets the criteria of SRP Section 3.5.1.6, no

additional design or procedural protection is required. Note that

the SRP criteria is only being used as one acceptable method of

evaluating risk. Use of this method is not a commitment to the SRP

and does not incorporate the SRP into our licensing basis.

Changes to the aircraft flight patterns and/or frequency

(probability) have no affect on the design or method of operating

equipment necessary to mitigate the consequences of previously

analyzed accidents. As was noted above, the aircraft hazard was

considered to be acceptable and, therefore, no additional design or

procedural protection is required for the plant. Since the aircraft

hazard is considered acceptable (where additional design features

are not required), it can be concluded that no action assumed to

occur within the accident analysis of CCNPP's Updated Final Safety

Analysis Report Chapter 14 will be degraded or prevented. Therefore,

it is concluded that the current calculated aircraft hazard will not

result in an increase of the consequences of an accident preciously

evaluated in the UFSAR.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

All possible malfunctions have been previously analyzed.

Aircraft hazard was addressed within the original design of the

plant. The frequency/probability of an aircraft crash was considered

to be so low that special design provisions to protect against

aircraft crashes did not have to be considered during construction

of CCNPP. The current calculated aircraft hazard is considered to

still be within SRP Section 3.5.1.6 guidelines. The possibility for

a malfunction of a different type than preciously evaluated in the

UFSAR is not created.

Aircraft accidents were considered within the original plant

design. The probability of an aircraft accident resulting in

radiological consequences greater than 10 CFR Part 100 exposure

guidelines is still considered to be below the level of

acceptability (per SRP Section 3.5.1.6) and no special design

provisions are required. Since an aircraft crash is not a design

basis concern, it is not plausible that the possibility of a new

accident is created that has not been previously evaluated in the

UFSAR. There are also no new challenges to safety-related equipment.

Therefore, the proposed change does not create the possibility

of a new or different type of accident from any accident previously

evaluated.

3. Would not involve a significant reduction in the margin of

safety.

The probability of an aircraft crash affecting the plant, at the

time of original licensing and construction, was so low that no

special design provisions were needed in the plant for such an

event. Since aircraft hazards did not have to be considered within

the design of the plant, no margin of safety was required or

established for such a hazard. All of the plant equipment and

initial condition assumptions stipulated within the UFSAR Chapter 14

accident analysis would not be affected by such an event.

The calculated probability of an aircraft accident resulting in

radiological consequences greater than 10 CFR Part 100 exposure

guidelines, based on today's aircraft hazard, is considered to be

below the 1.0 x 10-7 per year stipulated within SRP

Section 3.5.1.6. Therefore, there is still no need for special

design provisions within the plant to guard against such an event.

All of the plant equipment and initial condition assumptions

stipulated within the UFSAR Chapter 14 accident analysis remain

unchanged. The plant will continue to operate in such a manner that

will ensure acceptable levels of protection for the health and

safety of the public.

Therefore, this proposed change does not significantly reduce

the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Sheri R. Peterson.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: November 23, 1999

Description of amendments request: The requested amendments would

change Technical Specification (TS) 5.5.7.c.1, ``Ventilation Filter

Testing.'' The testing criteria would be changed consistent with the

NRC request in Generic Letter 99-02, ``Laboratory Testing of Nuclear-

Grade Activated Charcoal.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment revises TS 5.5.7.c.1 to require testing

of the SGT [Standby Gas Treatment] system charcoal in accordance

with American Society for Testing and Materials (ASTM) D3803-1989,

``Standard Test Method for Nuclear-Grade Activated Carbon.'' Per the

existing TSs, the SGT system charcoal must meet an acceptance

criteria of 70%. CP&L performs this testing in

accordance with the criteria of Regulatory Position C.6.a of

Regulatory Guide 1.52, Revision 1, 1976, ``Design, Testing, and

Maintenance Criteria for Engineered Safety Feature Atmosphere

Cleanup System Air Filtration and Adsorption Units of Light-Water-

Cooled Nuclear Power Plants.'' As stated in Updated Final Safety

Analysis Report, Section 6.5.1.1, the purpose of the SGT system,

along with that of the primary and secondary containment, is to

mitigate accident consequences. It is not associated with any

initiating events and, therefore, cannot affect the probability of

any accident.

ASTM D3803-1989 is an industry accepted standard for charcoal

filter testing. The conditions employed by this standard were

selected to approximate operating or accident conditions of a

nuclear reactor which would severely reduce the performance of

activated carbons. The key difference associated with the two

testing protocols is the testing temperature. Specifically, testing

to a challenge temperature of 30 deg.C per ASTM D3803-1989 versus

80 deg.C per Regulatory

[[Page 73087]]

Guide 1.52 results in a much more stringent test. Testing at a

higher temperature tends to eliminate impurities and moisture from

the sample. This creates the possibility of the charcoal achieving a

slightly higher efficiency than actual. Other parameter changes will

not significantly affect charcoal test performance and will result

in more accurate and reproducible test results.

The proposed TS change also includes a requirement that the test

be performed with a face velocity of 61 fpm. A single BSEP SGT

system train operates at a maximum flow rate of 4200 scfm which

corresponds to a face velocity of 61 fpm. In accordance with Generic

Letter (GL) 99-02, this requirement has been included in TS

5.5.7.c.1.

As recommended by GL 99-02, the proposed amendment incorporates

a safety factor of 2 into the allowed methyl iodide penetration

limit. The existing TS 5.5.7.c.1 acceptance criteria of 99% does not

account for a safety factor. In previous testing, CP&L has applied

the safety factor provided by Regulatory Guide 1.52, Revision 1,

1976, to the laboratory testing results to ensure proper charcoal

performance. The proposed changes to TS 5.5.7.c.1 require that

charcoal samples, tested in accordance with the methodology of ASTM

D3803-1989, show the methyl iodide penetration to be eff 0.95) during core alterations and that

uniform boron concentration is maintained for reactivity control in

the water volume having direct access with the reactor vessel. The

proposed change in the definition of core alterations will allow

``non-reactive'' components, such as cameras, lights, tools, movable

incore detector thimbles, etc., to be moved or manipulated in the

vessel, with fuel in the vessel and the vessel head removed, without

constituting a core alteration. This is acceptable because these

types of components will have negligible effect on core reactivity,

and will not affect reactor coolant system boron concentration.

Therefore, operations using these types of components will not

adversely affect Keff or the shutdown margin.

Additionally, reactor subcriticality status is continuously

monitored in the control room during Operating Mode 6, as specified

in Specification 3/4.9.2, ``Instrumentation''. Thus, there will be

no reduction in a margin of safety resulting from the proposed

change.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

request for amendments involves no significant hazards consideration.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: October 14, 1999 (TS 99-12).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79

(Unit 2) by revising the Technical Specification (TS) surveillance

requirements for steam generator tube integrity by incorporating an

alternate repair criteria for axial primary water stress corrosion

cracking at dented tube support plate intersections.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), has provided its

analysis of the issue of no significant hazards consideration, which is

presented below:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of Sequoyah Units 1 and 2, in accordance with the

proposed license amendment, does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Examination of crack morphology for primary water stress

corrosion cracking (PWSCC) at dented intersections has been found to

show one or two microcracks well aligned with only a few uncorroded

ligaments and little or no other inside diameter axial cracking at

the intersection. This relatively simple morphology is conducive to

obtaining good accuracy in Non-destructive Examination (NDE) sizing

of these indications. Accordingly, alternate repair criteria is

established based on crack length and average and maximum depth

within the thickness of the tube support plate (TSP) or limited

extension outside the thickness of the TSP.

The application of the alternate repair criteria (ARC) requires

a condition monitoring assessment. If all indications satisfy the

structural limits with regard to bounding lengths and average

depths, the condition monitoring burst pressure requirements are

satisfied.

In addition, an operational assessment is performed to determine

the length/depth repair bases. The crack profiles are projected to

the end of the operating cycle for comparison with acceptance limits

(i.e., length limit and average depth limit). Burst pressures are

calculated from the depth profiles by searching the total crack

length for the partial length that results in the lowest burst

pressure. Because the burst pressure can be lower than that for the

longest acceptable crack length at its average depth, a fixed repair

limit is not established. The repair bases is obtained by projecting

the crack profile to the end of the next operating cycle and

determining if the burst pressure for the projected profile meets

the burst pressure margin requirements defined by [Westinghouse

Topical Report] WCAP-15128, Revision 1, dated August 1999. If the

projected end-of-cycle (EOC) burst margin requirements are

satisfied, the indication is left in service. Thus, the repair limit

relative to length and average depth assures that the operational

assessment requirements are satisfied.

Crack length limits are established in the WCAP to assure that

crack extension and growth outside of the TSP provides adequate

margin against burst for the free-span crack (i.e., 3DPNO

burst capability is maintained) in addition to the total crack

length. A repair limit is also established in the WCAP for maximum

depth to provide a high confidence that the indication will not

progress through the wall at the end of an operating cycle.

Based on the above, the proposed amendment does not result in

any increase in the probability or consequences of an accident

previously evaluated within the Sequoyah FSAR [Final Safety Analysis

Report].

B. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Implementation of the proposed S/G [steam generator] tube ARC

does not introduce any significant changes to the plant design

basis. A single or multiple tube rupture event would not be expected

in a S/G in which the plugging criteria has been applied. Both

condition monitoring and operational assessments are completed as

part of the implementation of ARC to determine that structural and

leakage margin exists prior to returning S/Gs to service following

inspections. If the condition monitoring requirements are not

satisfied for burst or leakage, the causal factors for EOC

indications exceeding the expected values will be evaluated. The

methodology and application of this ARC will continue to ensure that

tube integrity is maintained during all plant conditions consistent

with the requirements of draft RG [Regulatory Guide] 1.121 and

Revision 1 of RG 1.83.

A S/G tube rupture event is one of a number of design basis

accidents that are analyzed as part of a plant's licensing basis. In

the analysis of a S/G tube rupture event, a bounding primary-to-

secondary leakage rate equal to the operational leakage limits in

the TSs, plus the leak rate associated with the double ended rupture

of a single tube, is

[[Page 73101]]

assumed. For other design basis accidents such as a main steam line

break and loss of alternating current power, the tubes are assumed

to retain their structural integrity and exhibit primary-to-

secondary leakage within the limits assumed in Final Safety Analysis

Report (FSAR) accident analyses. The proposed ARC does not result in

an accident leakage rate in excess of that assumed or calculated in

SQN's current accident analyses.

Even under severe accident conditions, the potential for

significant leakage would be expected to be small and not

significantly different than for other degradation mechanisms

repaired to 40 percent depth limits. It is concluded that

application of the proposed ARC for PWSCC at dented TSP locations

results in a negligible difference from current 40-percent repair

limits.

TVA continues to implement a maximum operating condition leak

rate limit of 150 gallons per day (0.1 gallons per minute) per S/G

to preclude the potential for excessive leakage during all plant

conditions.

The possibility of a new or different kind of accident from any

previously evaluated is not created because S/G tube integrity is

maintained by inservice inspection and effective primary-to-

secondary leakage monitoring.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

Tube repair limits provide reasonable assurance that tubes

accepted for continued service without plugging or repair will

exhibit adequate tube structural and leakage integrity during

subsequent plant operation. The implementation of the proposed ARC

is demonstrated to maintain S/G tube integrity consistent with the

criteria of draft NRC Regulatory Guide 1.121. The guidelines of RG

1.121 describe a method acceptable to the NRC staff for meeting

General Design Criteria (GDC) 2, 4, 14, 15, 31, and 32 by ensuring

the probability or the consequences of S/G tube rupture remain

within acceptable limits. This is accomplished by determining the

limiting conditions of degradation of S/G tubing, for which tubes

with unacceptable cracking should be removed from service.

Upon implementation of the proposed ARC, even under the worst-

case conditions, the occurrence of PWSCC at the tube support plate

elevations is not expected to lead to a S/G rupture event during

normal or faulted plant conditions. All tubes are shown to retain

the margins of safety against burst consistent with the safety

factor margins implicit in the stress limit criteria of Section III

of the American Society of Mechanical Engineers [ASME] Code, for all

service loading conditions. In addition, all tubes have been shown

to retain a margin of safety against gross failure or burst

consistent with the stress limits of [Paragraph] NB-3225 of Section

III of the ASME Code under postulated accident conditions concurrent

with a safe shutdown earthquake.

In addressing the combined effects of loss-of-coolant accident

plus safe shutdown earthquake on the S/G component (as required by

GDC 2), it has been determined that tube collapse will not occur in

the Sequoyah S/Gs. This analysis is discussed in WCAP 13990, dated

May 1994. No tubes are excluded from the application of the proposed

ARC.

Based on the above, it is concluded that the proposed license

amendment request does not result in a significant reduction in

margin with respect to the plant safety analyses as defined in the

FSAR or TSs.

The NRC has reviewed the licensee's analysis and, based on this

review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Section Chief: Richard P. Correia.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: November 8, 1999.

Brief description of amendments: The proposed amendments would

change Technical Specification 5.5.11, ``Ventilation Filter Testing

Program (VFTP)'' to include the requirement for laboratory testing of

Engineered Safety Feature (ESF) Ventilation System charcoal samples per

American Society for Testing and Materials (ASTM) D3803-1989 and the

application of a safety factor of 2.0 to the charcoal filter efficiency

assumed in the plant design-basis dose analyses.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Do the proposed changes involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed changes only involve the laboratory testing

methodology performed on activated charcoal to help determine

whether the charcoal in the filtration units can remain in place or

[if it] require[s] replacement.

Generic Letter 99-02 intends to standardize the way nuclear-

grade activated charcoal is tested throughout the industry in order

to provide conservative filtration results as well as uniform and

repeatable tests. The purpose is to ensure the filtration systems

protect the Operators in the Control Room (GDC [General Design

Criterion] 19) as well as the public (10CFR100), in the event of a

radiological accident scenario.

The charcoal adsorber sample laboratory testing per ASTM D3803-

1989 is more stringent than the current testing practice and more

accurately demonstrates the required performance of the adsorbers

following a design ba[s]is LOCA [loss of coolant accident]. No

Licensing Basis Accidents or mitigation capability will be affected

by incorporation of these changes.

Therefore, this change will not result in a significant increase

in the probability or consequences of an accident previously

evaluated.

(2) Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

Plant procedures are only altered to the extent that the revised

specification will allow different reference standards for testing

activated charcoal. These changes ensure continued support of the

safety related ESF filtration equipment and do not affect their

failure or failure modes.

Therefore, this change will not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

(3) Do the proposed changes involve a significant reduction in a

margin of safety?

None of the changes being proposed alter the environmental

conditions maintained in the areas supported by the ESF filtration

systems during normal operations and following an accident. Also

these changes will not cause an increase in radiological releases

through the Primary Plant Ventilation Exhaust System. As a result,

the margin of safety for these functions remains the same.

Therefore, this change does not involve a significant reduction in

a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: December 3, 1999 (ULNRC-04158).

Description of amendment request: The proposed amendment requested

changes to Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and

Temperature Limits Report (PTLR),'' of the improved Technical

Specifications (ITS) that were issued on May 28, 1999, in Amendment No.

133. The current Technical Specifications (CTS) remain in effect until

the ITS are implemented on or before April 30, 2000. The proposed

changes to the ITS would approve the use of the PTLR by the licensee to

make changes to the plant pressure temperature limits and low

temperature overpressure protection

[[Page 73102]]

limits without prior NRC staff approval in accordance with Generic

Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and

Low Temperature Overpressure Protection System Limits,'' dated January

31, 1996. The proposed changes are: (1) Add the word criticality to ITS

Subsection 5.6.6.a as one of the reactor conditions for which RCS

pressure and temperature limits will be determined, (2) add the phrase

``and COMS PORV,'' where COMS PORV stands for cold overpressure

mitigation system power operated relief valve, to the the introductory

paragraph of ITS subsection 5.6.6.b to show that the analytical methods

listed in the subsection are also for the COMS PORV, and (3) replace

the two documents listed in ITS subsection 5.6.6.b by the reference to

the future NRC letter that approves the use of the PTLR and the

Westinghouse Topical Report, WCAP-14040-NP-A, Revision 2, ``Methodology

Used to Develop Cold Overpressure Mitigating System Setpoints and RCS

Heatup and Cooldown Limit Curves,'' dated January 1996, that provides

the methodology that will be used by the licensee in using the PTLR

report. The current plant pressure temperature limits and low

temperature overpressure protection limits are in the CTS and were

approved in Amendment No. 124, which was issued April 2, 1998.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change submits the PTLR, which contains the

relocated CTS heatup and cooldown, and COMS PORV limits and the

methodology used to calculate them, and the added references into

ITS 5.6.6. The proposed change is administrative in nature since it

is a movement of information from the CTS to a licensee controlled

document, and has prior NRC staff approval. The PTLR contains the

limit curves and the ITS requires more restrictive actions to be

taken when the limiting conditions for operation are not met than is

currently required by the CTS. The heatup and cooldown, and COMS

PORV limits within the PTLR will be implemented and controlled per

Callaway Plant programs and procedures and changes to the PTLR will

be performed per requirements of 10 CFR 50.59 to ensure that change

to these limits in the future will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

As stated earlier, the movement of the heatup and cooldown, and

COMS PORV limits from the CTS to the PTLR has no influence or

impact, nor does it contribute in any way to the probability or

consequences of an accident. No safety-related equipment, safety

function, or plant operations will be altered as a result of this

proposed change. The proposed change is administrative in nature

since it is a movement of requirements from the CTS to a licensee

controlled document, the PTLR, and the change adds references into

the ITS incorporating the licensee controlled document. Therefore,

the possibility of a new or different kind of accident from any

accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not affect the acceptance criteria for

an analyzed event. The margin of safety presently provided by the

CTS remains unchanged. There will be no effect on the manner in

which safety limits or limiting safety system settings are

determined nor will there be any effect on those plant systems

necessary to assure the accomplishment of protective functions.

Therefore, the proposed change is administrative in nature and does

not impact the operation of Callaway Plant in a manner that involves

a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Section Chief: Stephen Dembek.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: November 5, 1999, as supplemented on

December 3, 1999.

Description of amendment request: This proposed change revises the

applicability for the reactor power distribution limits and the Average

Power Range Monitor (APRM) gain adjustments. The applicability is

proposed to be revised to operation at 25% Rated Thermal

Power (RTP).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change does not involve an increase in the

probability or consequences of an accident previously evaluated

because the revisions standardize and make consistent the

applicability and actions for the reactor power distribution limits

in the current Technical Specifications. Since reactor operation

with these revised Specifications is fundamentally unchanged, no

design or analytical acceptance criteria will be exceeded. As such,

this change does not impact initiators of analyzed events or assumed

mitigation of accident or transient events. The structural and

functional integrity of plant systems is unaffected. Therefore, the

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

The proposed change does not affect any parameters or conditions

that could contribute to the initiation of any accident. No new

accident modes are created. No safety-related equipment or safety

functions are altered as a result of these changes. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

At thermal power levels 25% RTP and has no effect on

any safety analysis assumption or initial condition. Thus, the

margin of safety required for safety analyses [is] maintained.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

[[Page 73103]]

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of amendment request: November 15, 1999 (TSCR 202).

Description of amendment request: The proposed amendments would

change the Technical Specifications (TSs) in order to extend the

required frequency of the control rod exercise test (TS 15.4.1, Table

15.4.1-2, Item 10) from the current frequency of every 2 weeks to

quarterly.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not result in a significant increase in

the probability or consequences of an accident previously evaluated.

Relaxing the frequency of performance for a surveillance does

not result in any hardware changes, nor does it significantly

increase the probability of occurrence for initiation of any

analyzed events since the function of the equipment has remained

unchanged. The proposed frequency has been determined to be adequate

based on industry operating data as supported by the conclusions

reached in NUREG 1366 and NRC GL [Generic Letter] 93-05.

Surveillance tests are intended to provide assurance of

continued component operability. The frequency of performance of a

surveillance does not significantly increase the consequences of an

accident, as a change in frequency does not change the response of

the equipment in performing its specified function (i.e. the overall

functional capabilities of the rod control system will not be

modified). Increasing the interval of control rod exercise testing

will reduce the possibility of inadvertent testing related [to]

reactor trips and dropped rods, and resulting in fewer challenges to

safety systems and resultant plant transients.

This change does not involve a significant increase in the

consequences of an accident or event previously evaluated because

the source term, containment isolation or radiological releases are

not being changed by the proposed revision. Existing system and

component redundancy and operation is not being changed by the

proposed change. The assumptions used in evaluating the radiological

consequences in the PBNP Final Safety Analysis Report are not

invalidated. Therefore, this change does not affect the consequences

of previously evaluated accidents.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

This change does not introduce nor increase the number of

failure mechanisms of a new or different type of accident than those

previously evaluated since there are no physical changes being made

to the facility. The design and design basis of the facility remain

unchanged. The plant safety analyses remain unchanged. All equipment

important to safety will continue to operate as designed. Component

integrity is not challenged. The changes do not result in any event

previously deemed incredible being made credible. The changes do not

result in more adverse conditions nor result in any increase in

challenges to safety systems. Therefore, operation of the Point

Beach Nuclear Plant in accordance with the proposed amendment will

not create the possibility of a new or different type of accident

from any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in

a margin of safety.

The proposed change does not involve a significant reduction in

the margin of safety because existing component redundancy is not

being changed by this proposed change. There are no changes to

initial conditions contributing to accident severity or

consequences. The proposed surveillance frequency, as supported by

past test results, continues to provide the required assurance of

operability, such that safety margins established through the design

and facility license, including the Technical Specifications, remain

unchanged. Therefore, there are no significant reductions in a

margin of safety introduced by this proposed amendment.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Claudia M. Craig.

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notice was previously published as a separate

individual notice. The notice content was the same as above. It was

published as an individual notice either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. It is repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna

Nuclear Power Plant, Wayne County, New York

Date of application for amendment: October 20, 1999.

Brief description of amendment: The amendment changed the footnote

to the Improved Technical Specifications associated with the Design

Features Fuel Storage Specification 4.3.1.1.b which required that 2300

ppm boron be maintained in the Spent Fuel Pool.

Date of publication of individual notice in Federal Register:

November 19, 1999 (64 FR 63346).

Expiration date of individual notice: December 20, 1999.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

[[Page 73104]]

Street, NW., Washington, DC, and electronically from the ADAMS Public

Library component on the NRC Web site, http://www.nrc.gov (the

Electronic Reading Room).

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.

1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: September 14, 1999.

Brief description of amendments: The amendments approve the

administrative changes to PVNGS TS 5.5.2, Primary Coolant Sources

Outside Containment, to delete the references to the post-accident

sampling return piping of the radioactive waste gas system and the

liquid radwaste system, and TS 5.6.2, Annual Radiological Environmental

Operating Report, to delete the administrative requirement to include

in the report certain TLD [thermoluminescence dosimeter] results that

are no longer available.

Date of issuance: November 24, 1999.

Effective date: November 24, 1999, to be implemented within 60

days.

Amendment Nos.: Unit 1--122, Unit 2--121, Unit 3--121.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 20, 1999 (64 FR

56528).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 24, 1999.

No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of application for amendment: October 21, 1999.

Brief description of amendment: This amendment revises Technical

Specifications (TS) for the Shearon Harris Nuclear Power Plant by

implementing selected improvements described in NRC Generic Letter (GL)

93-05, ``Line-Item Technical Specifications To Reduce Surveillance

Requirements For Testing During Power Operation,'' dated September 27,

1993.

Date of issuance: December 17, 1999.

Effective date: December 17, 1999.

Amendment No: 93.

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 17, 1999 (64

FR 62705).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 17, 1999.

No significant hazards consideration comments received: No.

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz

Mill, Pennsylvania

Date of application for amendment: September 15, 1999, as

supplemented on October 4, 1999.

Brief description of amendment: This amendment changes the

decommissioning Technical Specifications dealing with controls for

ingress, egress, and equipment removal from containment.

Date of issuance: December 7, 1999.

Effective Date: December 7, 1999.

Amendment No: 11.

Facility License No. TR-2: This amendment changes the

decommissioning Technical Specifications.

Date of initial notice in Federal Register: November 3, 1999 (64 FR

59798).

The Commission has issued a Safety Evaluation for this amendment

dated December 7, 1999.

No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Inc., Docket No. 50-003,

Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York

Date of application for amendment: July 20, 1999.

Brief description of amendment: The amendment would revise the

Technical Specifications to change the senior license requirements for

the Operations Manager.

Date of issuance: December 15, 1999.

Effective date: As of the date of issuance and shall be implemented

within 90 days from the date of issuance.

Amendment No: 46.

Facility Operating License No. DPR-5: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 2, 1999 (64

FR 49027).

The July 20, 1999, letter providing clarifying information that did

not change the scope of the original application and proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated December 15, 1999.

No significant hazards consideration comments received: No.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of application for amendments: May 5, 1999, as supplemented

June 22 and July 30, 1999.

Brief description of amendments: These amendments conform the

licenses to reflect the transfer of Operating Licenses Nos. DPR-66 and

NPF-73 for the Beaver Valley Power Station Unit Nos. 1 and 2, to the

extent held by Duquesne Light Company (DLC) to the Pennsylvania Power

Company, and the operating authority under the licenses from DLC to

FirstEnergy Nuclear Operating Company as previously approved by an

Order dated September 30, 1999.

Date of issuance: December 3, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment Nos.: 226 and 104.

Facility Operating License Nos. DPR-66 and NPF-73: These amendments

revised the Operating Licenses.

Date of initial notice in Federal Register: June 14, 1999 (64 FR

31880).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated September 30, 1999. The June 22 and July

30, 1999, supplements were within the scope of the initial application

as originally noticed.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: September 14, 1999.

Brief description of amendment: This amendment eliminates License

Condition 2.C.10 of the Operating License regarding controls over the

containment air locks during plant outages and modifies License

Condition 2.F of the Operating License regarding reporting requirements

for violations of the Technical Specifications and the Environmental

Protection Plan.

Date of issuance: December 15, 1999.

Effective date: December 15, 1999.

Amendment No.: 109.

Facility Operating License No. NPF-58: This amendment revised the

Operating License.

Date of initial notice in Federal Register: November 3, 1999 (64 FR

59803).

The Commission's related evaluation of the amendment is contained

in a

[[Page 73105]]

Safety Evaluation dated December 15, 1999.

No significant hazards consideration comments received: No.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County, Pennsylvania

Date of application for amendment: June 29, 1999, as supplemented

August 27, October 29, and November 3, 1999.

Brief description of amendment: The amendment clarifies the

authority to possess certain types of radioactive materials and

components at either Unit 1 or Unit 2. Following the transfer of the

Three Mile Island, Unit 1 (TMI-1), operating license to AmerGen, these

items, under the amendment, may continue to be moved between the TMI-1

and TMI-2 units as they currently are.

Date of issuance: December 9, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 217.

Facility Operating License No. DPR-50: Amendment revised the

License.

Date of initial notice in Federal Register: July 12, 1999 (64 FR

37572). The August 27, October 29, and November 3, 1999, letters

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination or expand

the amendment beyond the scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 9, 1999.

No significant hazards consideration comments received: No.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear

Station, Unit 2, (TMI-2) Middletown, Pennsylvania

Date of application for amendment: June 29, 1999, as supplemented

by letters dated August 27, October 29, and November 3, 1999.

Brief description of amendment: The amendment adds a provision to

the license conditions to ensure that the storage of certain types of

radioactive materials and components at Three Mile Island (TMI), Unit

2, pursuant to the TMI, Unit 1 license, does not result in a source

term that would exceed the limits in the TMI, Unit 2 Post-Defueling

Monitored Storage Safety Analysis Report.

Date of issuance: December 14, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days.

Amendment No.: 53.

Facility Operating License No. DPR-73: Amendment revised the

License.

Date of initial notice in Federal Register: July 12, 1999 (64 FR

37572). The August 27, October 29, and November 3, 1999, supplements

provided clarifying information that did not change the initial

proposed no significant hazards consideration determination or expand

the amendment beyond the scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 14, 1999.

No significant hazards consideration comments received: No.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment: July 23, 1999, as supplemented

July 30, August 9, August 20, October 7, and October 11, 1999.

Brief description of amendment: The amendment replaces references

to Illinois Power Company in the Operating License with references to

AmerGen Energy Company, LLC, to reflect the transfer of the license as

approved by an Order dated November 24, 1999.

Date of issuance: December 15, 1999.

Effective date: December 15, 1999.

Amendment No.: 123.

Facility Operating License No. NPF-62: The amendment revised the

Operating License.

Date of initial notice in Federal Register: August 19, 1999 (64 FR

45290).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 24, 1999.

Comments received: Yes. Comments received from The Environmental

Law and Policy Center of the Midwest were addressed in the staff's

safety evaluation.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: September 23, 1999, as

supplemented October 11 and November 10, 1999.

Brief description of amendments: The amendments provide approval to

move steam generator sections through the auxiliary building and to

disengage crane travel interlocks, and provide relief from performance

of Technical Specification Surveillance Requirement 4.9.7.1. The loads

to be moved are in support of the Unit 1 Steam Generator Replacement

Project.

Date of issuance: December 7, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment Nos.: 233 and 216.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 26, 1999 (64 FR

57665). The October 11, 1999, submittal provided corrected TS pages.

The November 10, 1999, submittal was in response to a NRC request for

additional information dated October 26, 1999, and provided clarifying

information to the original submittal. This information was within the

scope of the original Federal Register notice and did not change the

staff's initial proposed no significant hazards considerations

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 7, 1999.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: October 1, 1999, as

supplemented November 19, 1999.

Brief description of amendments: The amendments involve the

resolution of an unreviewed safety question related to certain small-

break loss-of-coolant accident scenarios for which there may not be

sufficient containment recirculation sump water inventory to support

continued operation of the emergency core cooling system and

containment spray system pumps during and following switchover to cold

leg recirculation. Resolution of this issue consists of a combination

of physical plant modifications, new analyses of containment

recirculation sump inventory, and resultant changes to the accident

analyses to ensure sufficient water inventory in the containment

recirculation sump. The amendments would also change the Technical

Specifications dealing with the refueling water storage tank inventory

and temperature, the required amount of ice in each ice basket in the

containment, and the delay to start the containment air recirculation/

hydrogen skimmer fans.

Date of issuance: December 13, 1999.

Effective date: As of the date of issuance and shall be implemented

within 60 days.

Amendment Nos.: 234 and 217.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

[[Page 73106]]

Date of initial notice in Federal Register: October 29, 1999 (64 FR

58458).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 13, 1999.

No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: February 12, 1999.

Brief description of amendment: The amendment changes the Technical

Specifications to (1) allow reactor vessel hydrostatic and leakage

tests when reactor coolant temperature is above 212 deg.F without

maintaining primary containment integrity and (2) establish a limit and

a surveillance requirement on reactor coolant activity when reactor

coolant temperature is above 212 deg.F, the reactor is not critical,

and primary containment has not been established.

Date of issuance: November 24, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment No.: 107.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 24, 1999 (64 FR

14283).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated November 24, 1999.

No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of application for amendment: September 30, 1999.

Brief description of amendment: The amendment changes the Technical

Specification surveillance periodicity requirements for the control

room emergency filtration system.

Date of issuance: December 8, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment No.: 108.

Facility Operating License No. DPR-22. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 3, 1999 (64 FR

59805).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 8, 1999.

No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company,Delmarva

Power and Light Company, and Atlantic City Electric Company, Docket No.

50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,

Pennsylvania

Date of application for amendment: March 1, 1999, as supplemented

June 14, October 1 and October 6, 1999.

Brief description of amendment: The amendment supports the

installation of a digital Power Range Neutron Monitoring system and the

incorporation of the long-term thermal-hydraulic stability solution

hardware.

Date of issuance: October 14, 1999.

Effective date: Effective as of date of issuance and shall be

implemented prior to restart from the Peach Bottom Atomic Power

Station, Unit 3, October 1999 refueling outage.

Amendment No.: 234.

Facility Operating License No. DPR-56: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 2, 1999 (64 FR

29711). The June 14, October 1 and October 6, 1999, provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated October 14, 1999.

No significant hazards consideration comments received: No.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: April 6, 1999.

Brief description of amendment: The amendment changes the Technical

Specifications by removing the words ``three individual underground''

and ``underground'' from the limiting conditions for operation when

referring to the emergency diesel generator fuel oil storage tanks in

Sections 3.7.A.5 and 3.7.F.4.

Date of issuance: December 7, 1999.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 198.

Facility Operating License No. DPR-64: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 2, 1999 (64 FR

29713).

No significant hazards consideration comments received: No.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 7, 1999.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., (SNC) Docket Nos. 50-348 and

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Dates of amendments request: March 12, 1998, as supplemented by

letters of April 24, 1998, August 20, 1998, November 20, 1998, February

3, 1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999,

June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999,

September 15, 1999, and September 23, 1999.

Brief description of amendments: The amendments fully convert SNC's

Current TS (CTS) to Improved TS (ITS) based on NUREG-1431, ``Standard

Technical Specifications, Westinghouse Plants,'' Revision 1, of April

1995. The amendments add two new Additional Conditions to Appendix C of

the Unit 1 and Unit 2 Facility Operating Licenses. The first new

Additional Condition authorizes SNC to relocate certain CTS

requirements to SNC-controlled documents. The second new condition

addresses the schedule for performing new and revised ITS

surveillances.

Date of issuance: November 30, 1999.

Effective date: As of the date of issuance and shall be implemented

no later than March 31, 2000.

Amendment Nos.: 146 and 137.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments fully

convert SNC's CTS to ITS.

Dates of initial notices in Federal Register: May 25, 1999 (64 FR

28218) and August 25, 1999 (64 FR 46443). The supplemental letters

dated April 24, 1998, August 20, 1998, November 20, 1998, February 3,

1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999,

June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999,

September 15, 1999, and September 23, 1999, provided clarifying

information that did not change the initial proposed no significant

hazards consideration determinations.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated November 30, 1999.

No significant hazards consideration comments received: No.

[[Page 73107]]

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424

and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke

County, Georgia

Date of application for amendments: April 28, 1999.

Brief description of amendments: The amendments revised Vogtle's

operating licenses to allow the licensee to establish containment

hydrogen monitoring within 90 minutes of initiation of a safety

injection following a loss-of-coolant accident, compared to the current

30 minute requirement.

Date of issuance: December 8, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: 110 and 88.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Operating Licenses.

Date of initial notice in Federal Register: August 11, 1999 (64 FR

43779).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 8, 1999.

No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: July 28, 1998, as supplemented by

letters dated May 31 and October 21 (2 letters), 1999.

Brief description of amendments: The amendments authorize the

revision of the South Texas Project updated final safety analysis

report (UFSAR) to allow the use of operator action to reduce the steam

generator power-operated relief valve setpoint consistent with the

revised small-break loss-of-coolant accident analysis for the

replacement Delta 94 SGs.

Date of issuance: December 14, 1999.

Effective date: December 14, 1999. Revisions will be incorporated

into the next UFSAR update in accordance with the schedule in 10 CFR

50.71(e).

Amendment Nos.: Unit 1--119, Unit 2--107.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

authorize revision of the UFSAR.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48268).

The May 31 and October 21 (2 letters), 1999, supplements provided

additional clarifying information. One of the October 21, 1999,

supplements also provided a revised UFSAR pages. This information was

within the scope of the original application and Federal Register

notice and did not change the staff's initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 14, 1999.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County,

Alabama

Date of application for amendments: September 30, 1999.

Description of amendment request: The amendments revise the

operating licenses to remove license conditions that have become

outdated, are no longer applicable, or are redundant, and to

consolidate license conditions which currently exist in two locations

in each units license.

Date of issuance: December 16, 1999.

Effective date: December 16, 1999.

Amendment Nos.: 237, 262, and 222.

Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:

Amendments revised the licenses.

Date of initial notice in Federal Register: November 3, 1999 (64 FR

59807).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1999.

No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: February 27, 1998, as supplemented by

letters dated June 10, 1998, and October 22, 1999.

Brief description of amendments: The amendments change the

refueling water storage tank (RWST) low-low level setpoints in

Technical Specification Table 3.3.2-1, ``Engineered Safety Feature

Actuation System Instrumentation,'' to increase the volume of water

available to containment spray pumps when the containment spray system

switches to the recirculation mode of operation.

Date of issuance: December 8, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: 73 and 73.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 15, 1998 (63 FR

38205). The October 22, 1999, supplement provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination or expand the scope of the

application beyond the scope described in the initial notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 8, 1999.

No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 18, 1999.

Brief description of amendment: The amendment revises the reactor

core spiral reloading pattern such that it begins around a source range

monitor. The offloading pattern is the reverse sequence.

Date of Issuance: December 14, 1999.

Effective date: As of its date of issuance, and shall be

implemented within 30 days.

Amendment No.: 181.

Facility Operating License No. DPR-28: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 8, 1999 (64

FR 48867).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated December 14, 1999.

No significant hazards consideration comments received: No.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: September 23, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications (TSs) by deleting the test requirements for

snubbers from the TSs. These requirements are already included in the

Point Beach Nuclear Plant In-Service Inspection Program.

Date of issuance: December 6, 1999.

Effective date: As of the date of issuance and shall be implemented

within 45 days.

Amendment Nos.: 191 and 196.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

[[Page 73108]]

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71977).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 6, 1999.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: December 29, 1998, as supplemented by

letters dated July 29 and October 21, 1999.

Brief description of amendment: The amendment revised (1) the

reactor coolant system (RCS) heatup and cooldown limit curves in

Figures 3.4-2 and 3.4-3 and cold overpressure mitigation system power-

operated relief valve setpoint limit curve in Figure 3.4-4 of the

current TSs, and (2) the list of references in Section 5.6.6 on the RCS

pressure temperature limits report (PTLR) in the improved TSs. The

improved TSs were issued in Amendment No. 123, dated March 31, 1999, to

replace the current TSs, but have not yet been implemented. The

revision to Section 5.6.6 of the improved TSs replaced the previous

references to NRC documents giving criteria for the above limit curves

in the current TSs by the references to (1) the NRC letter of December

2, 1999, that approved the use of the PTLR of Generic Letter 96-03,

``Relocation of the Pressure Temperature Limit Curves and Low

Temperature Overpressure Protection System Limits,'' dated January 31,

1996, for WCGS, and (2) WCAP-14040-NP-A, ``Methodology Used to Develop

Cold Overpressure Mitigation System Setpoints and RCS Heatup and

Cooldown Limit Curves.'' The PTLR will provide the methodology for the

licensee to revise the heatup and cooldown and setpoint limit curves

for WCGS in the future without prior staff approval, after the improved

TSs are implemented and have replaced the current TSs. The improved TSs

are to be implemented by December 31, 1999.

Date of issuance: December 7, 1999.

Effective date: December 7, 1999, to be implemented by December 31,

1999.

Amendment No.: 130.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9023) and September 8, 1999 (64 FR 48869). The October 21, 1999,

supplemental letter provided additional clarifying information, did not

expand the scope of the application as originally noticed, and did not

change the staff's original proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 7, 1999.

No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: November 8, 1999.

Brief description of amendment: The amendment corrects 15 errors in

the improved Technical Specifications that was issued in Amendment No.

123 on March 31, 1999. In addition, four corrections to Table LG,

``Details Relocated from Current Technical Specifications [CTS],'' that

was attached to the safety evaluation dated March 31, 1999, issued with

Amendment No. 123 were made.

Date of issuance: December 16, 1999.

Effective date: December 16, 1999, to be implemented December 31,

1999.

Amendment No.: 131.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 16, 1999 (64

FR 62231).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1999.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 8th day of December 1999.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Deputy Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-33684 Filed 12-28-99; 8:45 am]

BILLING CODE 7590-01-P